The following pertains to the nuclear reactor arts, nuclear power arts, nuclear reactor safety arts, and related arts.
Existing nuclear power plants are typically light water thermal nuclear reactors of the boiling water reactor (BWR) or pressurized water reactor (PWR) designs. In such a reactor, a nuclear reactor core comprising fissile material (typically a uranium compound such as UO2 enriched in fissile 235U) is disposed in coolant (purified water) contained at an operational pressure and temperature in a reactor pressure vessel. A nuclear chain reaction involving fission of the fissile 235U generates heat in the nuclear reactor core which is transferred to the coolant. In a BWR design, the heat directly converts coolant to steam, and steam separator/dryer hardware contained in the reactor pressure vessel generates steam that is output via large-diameter piping to a turbine to generate electricity (in a nuclear power plant setting; more generally the output steam is used to perform other useful work). The condensed coolant from the turbine is fed back into the BWR pressure vessel via additional large-diameter piping. In a PWR design, the primary coolant remains in a liquid state (e.g. subcooled) and is piped via large-diameter piping to an external steam generator where heat from the (primary) reactor coolant converts (separate secondary) coolant to steam that in turn drives the turbine. The condensed coolant from the steam generator is fed back into the PWR pressure vessel via additional large-diameter piping.
In such designs, the reactor pressure vessel is relatively compact. It contains the reactor core and associated internals such as control rods, and (in the case of a BWR) the steam separator/dryer hardware, along with attached ancillary equipment such as control rod drive systems and valves. The nuclear reactor core is typically the heaviest component and it is located in the lower portion of the reactor pressure vessel so as to reduce likelihood of the core being uncovered in the event of a loss of coolant accident (LOCA). The large-diameter piping connecting the reactor pressure vessel with the coolant loop to the turbine (for a BWR) or steam generator (for a PWR) also provides structural support for the compact reactor pressure vessel.